ML13241A383

Issuance of Amendments Spent Fuel Pool Criticality Changes

ML13241A383
Person / Time
Site:	Prairie Island
Issue date:	08/29/2013
From:	Thomas Wengert Plant Licensing Branch III
To:	Jeffery Lynch Northern States Power Co
Wengert T
References
TAC ME6984, TAC ME6985
Download: ML13241A383 (36)

Text

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. OFFIGI~ USE ONl:Yp~~PRIE~RY INFORMATION UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555.0001 August 29, 2013 Jam~s E. L'jf\\ch

. Site Vice President Northem States Power Company - Minnesota Prairie Island Nuclear Generating Plant 1717 Wakonade Drive East Welch, MN 55089-9642

SUBJECT:

PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 AND 2 -ISSUANCE OF AMENDMENTS RE: SPENT FUEL POOL CRIllCALITY CHANGES (T AC NOS. ME6984 AND ME6985)

Dear Mr. Lynch:

The U.S. Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment NO..209 to Renewed Facility Operating License No. DPR-42 and Amendment No. 196 to Renewed Facility Operating License No. DPR-60 for the Prairie Island Nuclear Generating Plant, Units 1 and 2. respectively. The amendments consist of changes to the Technical Specifications (TSs) in response to your application dated August 19, 2011, supplemented by letters dated May 16, 2012. September 4,2012, February 8,2013, and July 17,2013.

The amendments revise TS 3.7.17. "Spent Fuel Pool Storqge,' and TS 4.3.1, ~Fuel Storage Criticality" to provide new spent fuel pool (SFP) loading restrictions that meet subcriticality for all postulated conditions. The TS changes correct non-conservatislT's in the SFP criticality analysis-of-record. The amendments also change the evaluation methodology used for the SFP criticality analysis.

I NOTICE: Endosure 4 to this letter contains Proprietary Information.

'I Upon separation from EnClosure 4, '

this letter is DECONTROLLED OFFICIAL'~ ONLY' PROPRIETARY INFORMATION'

OFFICIAL USE ONLY PROPRIETARY INFORMATION J. Lynch

- 2 The NRC has determined that the related Safety Evaluation (SE) contains proprietary information pursuant to litle 10 of the Code of Federal Regulations, Section 2.390, 'Public Inspections, Exemptions, Requests for Withholding." Proprietary information is indicated by text enclosed within double brackets. Accordingly. the NRC staff has also prepared a redacted

. publicly available, non-proprietary version of the SE. Copies of the proprietary and non proprietary versions of the SE ar~ enclosed. The Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Sincerely,

.-,=F-d~

Thomas J. Wengert, Senior Project Manager Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-282 and 50.:.306

Enclosures:

1. Amendment No. 209 to DPR-42

2. Amendment No. 196 to DPR-60

3. Non-Proprietary Safety Evaluation

4. Proprietary Safety Evaluation

'cc w/encls 1, 2, and 3: Distribution via ListSerV OFFICIAL.USE ONLY PROPRIETARY INFORMATION

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-4001 NORTHERN STATES POWER COMPANY - MINNESOTA DOCKET NO. 50-282 PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNIT 1*

AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 209 license No. DPR-42

1.

The U.S, Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Northern States Power Company, a Minnesota Corporation (NSPM. the licensee). dated August 19. 2011. as supplemented by letters dated May 16, 2012, September 4, 2012, February 8,2013, and July 17, 2013, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisipns of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is hereby 'amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of.

Renewed Facility Operating license No. DPR-42 is hereby amended to read as follows:

-2 Technical Sl2ecifications The Technical Specifications contained in Appendix A. as revised through Amendment No. 209, are hereby incorporated in the re.newed operating license.

NSPM shall operate the facility in accordance With the Technical Specifications:

3.

Implementation Requirements

. This license amendment is effective as of the date of its issuance and shall be implemented INithin 120 days. In conjunction INith implementation of the amendment, procedures will be revised to require an assessment of a fuel assembly's exposure to rodded power operation in the core prior to moving that fuel assembly into the spent fuel pool (SFP) storage racks. If an assembly experiences more than 100 megawatt day per metric ton uranium (MWd/MTU) of core average full-power rodded operation exposure, this exposure experienced while rodded will not be credited for determining the

. coefficients used to categorize fuel assemblies as described in WCAP-17400-P.

.~

FOR THE NUCLEAR REGULATORY COMMISSION Robert D. Carlson, Chief Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor* Regulation

Attachment:

Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: AI-lgust 29, 2013

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555..0001 NORTHERN STATES POWER COMPANY - MINNESOTA DOCKET NO. 50-306 PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNIT 2 AMENDMENT to RENEWED FACILITY OPERATING LICENSE Amendment No. 196 License No.'DPR-60

1.

The U.S, Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Northern States Power Company, a Minnesota Corporation (NSPM, the licensee). dated August 19. 2011, as supplemented by letters dated May 16i 2012, September 4, 2012, February 8,2013, and July 17, 2013, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations setforth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ji)that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and securitY or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

AccordinglV, the license is hereby amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-60 is her.eby amended to read as follows:

- 2 Technical Specifications The Technical Specifications contained in Appendix A. as revised through Amendment No. 196, are hereby incorporated in the renewed operating license.

NSPM shall operate the facility in accordance with the Technical Specifications.

3.

Implementation Requirements This license amendment is effective as of the date of its is.suance and shall be implemented within 120 days. In conjunction with implementation of the amendment, procedures will be revised to require an assessment of a fuel assembly's exposure to rodded power operation in the core prior to moving that fuel assembly into the spent fuel pool (SFP) storage racks. If an assembly experiences more than 100 megawatt day per metric ton uranium (MWd/MTU) of core average full-power rodded operation exposure, this exposure experienced while rodded will not be credited for determining the coefficients used to categorize fuel assemblies as described in WCAP-17400-P.

FOR THE NUCLEAR REGULATORY COMMISSION

~~

Robert D. Cartson. Chief Plant licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: August 29. 2013*

ATTACHMENT TO LICENSE AMENDMENT NOS. 209 AND 196 RENEWED FACILITY OPERATING LICENSE NOS. DPR-42 AND DPR-60 DOCKET NOS. 50-282 AND 50-306 Replace the following pages of the Renewed Facility Operating License Nos. DPR-42,and DPR-60 with the attached revised pages. The revised pages are identified by amendment '

number and contain marginal lines indicating the areas of change.

REMOVE INSERT DPR-42, License Page 3 DPR-42, LIcense Page 3 DPR-60, License Page 3 DPR-60, license Page 3 Replace the following pages of the Appendix A Technical SpeCifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal

. lines indicating the areas of change.

REMOVE INSERT 3.7.17-1 3.7.17-1 3.7.17-2 3.7.17-2 3.7.17-3 4.0-2 4.0-2 4.0-3 4.0-3 4.0-5 4.0-5

(,

4.0-6 4.0-6 4.0-7'

/'

4.0-7 4.0-8 4.0-8 4.0-9 4.0-10

-3 (3)

Pursuant to the Act and 10 CFR Parts 30, 40 and 70, NSPM to receive, possess, and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup. sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4)'

Pursuant to the Act and 10 CFR Parts 30, 40, and 70, NSPMto receive, possess and use in amounts as required any byproduct; source or special nuclear material without restriction to chemical or physical form, for sample analysis\\ or instrument and equipment ca\\ibration or associated with radioactive apparatus or components; (5)

Pursuant to the Act and 10 CFR Parts 30 and 70. NSPM to possess but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility; (6)

Pursuant to the Act and 10 CFR Parts 30 and 70, NSPM to transfer byproduct materials from other Job sites owned by NSPM for the purpose of volume reduction and. decontamination.

C.

This renewed operating license shall be deemed to contain and is subject to the conditions specified in the follOWing Commission regulations in 10 CFR Chapter I:

Part 20,. Section 30.34 of Part 30, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; is subject to all applicable provisions of the Act and to the rules, regulations. and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1 )

Maximum Power Level NSPM is authorized to operate the facility at steady state reactor core power levels not in eXCess of 1677 megawatts thermal.

. (2)

Technical Specifications The TeChnical Specifications contained in Appendix A. as revised through Amendment No. 209, are hereby incorporated in the renewed operating license. NSPM shall operate the facility in accordance with the Technical Specifications.

(3) physical Protection NSPM shall fully implement and maintain in effect all provisions, of the Commission-approved physical security, guard training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Misce"aneous Amendments and Search Requirements revisions to 10 CFR. 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR SO.54(p). The combined set of plans, which contains Renewed Operating License No. DPR-42 Amendment No. 209

-3 (3)

Pursuant to the Act and 10 CFR Parts 30. 40 and 70. NSPM to receive.

possess. and use at any time any byproduct. source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration. and as.

fission detectors in amounts as required;.

(4)

. Pursuant to the Act and 10 CFR Parts 30, 40, and 70, NSPM to receive, possess and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument and equipment calibration or associated with radioactive apparatus or components:

(5)

P.ursuant to the Act and 10 CFR Parts 30 and 70, NSPM to possess but not separate. such byproduct and special nuclear materials as may be produced by the operation of the facility; (6)

Pursuant to the Act and 10 CFR Parts 30 and 70, NSPM to transfer byproduct materials from other job sites owned by NSPM for the purposes of volume reduction and decontamination, C.

This renewed operating license shall be deemed to contain and is subjectJo the conditions speCified in the follOwing Commission regulations in 10 CFR Chapter I:

Part 20, Section 30.34 of Part 30, Sections 50.54 and 50.59 of Part 50; and Section 70.32 of Part 70; is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1)

Maximum Power Level NSPM is authorized to operate the facility at steady state reactor core power levels not in excess of 1677 megawatts thermal.

(2)

Technical SpeCifications

\\

The Technical Specifications contained in Appendix A, as revised through Amendment No. 196, are hereby incorporated in the renewed operating license.

NSPM shall operate the facility in accordance with the Technical Specifications.

(3)

Physical protection NSPM shaH fully implement and maintain in effect all provisions of the Commission-approved physical security, guard training.and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined set of plans, which contains Renewed Operating License No. DPR~60 Amendment No. 196

t.

'Spent Fuel Pool Storage 3.7.17 3.7 PLANT SYSTEMS

_"",.1 3.7.17 Spent Fuei Pool Storage LCO 3.7.17 Each fuel assembly, fuel insert, or hardware stored in the spent fuel pool shaJJ satisfy the loading restriciions of Specification 4.3.1.1.

. APPUCABll.1IY:

Whenever any fuel assembly is stored in the spentJueJ storage pool.

ACTIONS Lo an acceptable location.

CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of the LCO not met A.I -------------NOTE------------

LCO 3.0.3 is not applicabJe.

Initiate action to move the

, noncomplying fuel assembly Immediately Prairie Island'

,Unit 1-- Amendment No..J...§.8 ~ 209 Units 1 and 2

'3.7.17-1 Unit 2 - Amendment No. 149 ~ 196 I' i

Spent Fuel Pool Storage 3.7.17 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.17.1 Verify by administrative means the fuel assembly, Prior to storing fue! insert, or other hardware placed in the spent fuel or moving the storage racks is,stored in accordance with fuel assembly, Spedfication 4.3.1.1.

fuel insert, or other'hardware SR 3.7 J 7.2 Verify spent fuel pool inventory.

Within 7 days after completion of a spent fuel pool fuel handling Prairie Island Unit 1.

Amendment No. ~ m 2091 Units land 2 3.7.17-2 Unit 2 - Amendment No,. +49~ 19~

Design Features 4.0 4.0 DESIGN FEATURES 4.3 Fuel.Slorage 4.3.1 Criticality 4.3.1.1 The spent fuel storage racks are designed and shall be maintained*

with:

a.

Fuel as~emblies having a maximum U-235 enrichment of 5.0 weight percent;

b.

kerf < 1.0 if fully flooded with unborated water, which includes an aHowance for uncertainti.es as described in VSAR Section 1(U;

c.

kerf:5 0.95 if fully flooded with water borated to 400 ppm, '

which' inCludes an allowance for uncertainties as described in VSAR Section 10.2;

d.

A nominal 9.5 inch center to center distance between fuel assemblies placed in the fuel srorage racks;

e.

New or spent fuel assemblies, fuel inserts,' and hardware loaded in accordance with Figure 4.3.1-1.

Prairie Island Unit 1 - Amendment No. m m 209 !

L: nits 1 and 2 4.0-2 Vnit 2 - Amendment No. ~ +&+ 196.

Design Features 4.0 4.0 DESIGN FEATURES 4.3 Fuel Storage (continued) 4.3'.1.2 1he new fuel storage racks are designed and shall be maintained with:

a.

Fuel assemblies having a maximum U-235 enrichment of 5.0 weight percent; b:k"f[ $. 0.95 if fully flooded with unborated water, which includes an allowance for uncertainties as described in USAR Section 10.2;

c.

kerf $. 0.98 if accidentally filled with a Jow density moderator which resulted in optimum low density moderation conditions; and

d.

A.nominal 21 inch center to center distance between fuel assemblies placed in the storage racks.

4.3.2 Drainage The spent fuel storage pool is designed and shan be maintained to prevent inadvertent draining of the pool below eievation 727' 4" (Mean' Sea Level).

Prairie Island Unit 1 - Amendment No. f:i.8 m 209 Units 1 and 2 4.0-3 Unit 2 - Amendment No, M9.:t.3+196

4.0 Design Features Table 4.3.1*} (page I of I)

Fuel. Categories Ranked by Reactivhy FUEL CATEGOR Y RELAfiVE REACTIVITY High 2

3 4

5 6

Low 7

I Consolidmed Fuel NOles:

.1.

Fucl calegory is ranked by decreasing order of reactivily wilhout regard for nny reactivity*reducing mechanisms, e.g.,

Cmegory 2 is less reactive than c.iHi:gory I, clc. The more reaclive fuel categories require additional measures 10 be pl;u:cd on fuel placement i'n the Spent Fuel Poc>i (SFP} racks, e.s., more use of waler-filled cells or Rod Comrol Clusler Assemblies (RCCAs).

2.

Any higher-numbered fuel category (except Category 7) may be used in an array specifyinr, a lower*numbered fuel category.

3.

Category I is fuel up lO 5.0 weight percent U-235 enrichment and does nOt credit bumup.

4.

Category 7 i~ consolidated fuel stored in Consolidated Rod SlOrag" Canister~.

5, Categories 2 thrQugh 6 are determined from Tables 4.3.1-2 and 4.3.1*3.

Prairie Island Unit 1 - Amendment No. 209 Units 1 and 2 4.0-5 UniL 2..:. Amendment No. 196

Design Features 4.0 Table 4,3, I *2 (page I of I)

For Fuel Operated in Units I and 2 Cycles I *4

, Coefficienls to Calculate Ihe Minimum Required Fuel Assembly Bum~p (Bu) as a Function of Decay Time and Enrichmem (En)

FUEL DECAY TIME COEFFICIENTS CATEGORY Al I

Al A~

A.

I I

0,000

-0,722 14.272 0

-3U67 3

20 O.QOO

• 1.944 20.494

• 39.085 I

1 I

0 0,673

-8.24~

44.607

-56.428

5.

• 20 1.784

• 16.297 60.035

-64.713 i

• 10.246

, I

.47:457 0

1.097

• 56.456.

6 20 1.820 I

-15.656 56856

• 60.351 i

i Notes:

I.

AU relevant uncertainties an:: explicitly included in the crilicalilY analysis. For insIDnce, no additional allowance for bumup uncertainlY or enrichment uncertainty is required. For a fuel assembly 10 meet the requirements of a Fuel Category, Ihe assembly bumup musl exceed "minimum burnup~ (GWd/MTU) given by Ihe curve fil for Ihe assembly "decay time" and "inilial enrichmem". The specific minimum bumup required for each fuel assembly ;$ calculalcd from the following equ31ion for each increment of decay lime:

2.

Iniliu! enrichment (En) is Ihe nominal U*235 enrichment, Any enrichmeOi between 1.7 and 3.4 weight °percenl U*235 may be used. If tne com pilicd Bu value is negalive, zero snail be usedo 3,

Decay Time is in years. An assembly Wllh a cooling lime greater Ihan 20 year, musl U!'e 20 years. No extrapolation i, permilled.

4.

If Deeay Time value falls between increments of the laoble, the lower Decay Time value shall be used or a linear interpolation may be performed as follows: Compule the Bu value using tile coeffIcients assod:lIed wilh the Decay Time values Ihal brackellhe aClI.lul Decay Time, Interpolate between Bu values based on the increl1lem of Decay Time bel ween Ihe acwal Decay Time value and the computed Bu results.

5.

This table applieli (0 fuel assemblies Ihat were operated in Ihe core fur any period of time during Unit I or Unil 2 Cycles I through 40 Prairie Island Unit 1 Amendment No. 209 Units 1 and 2 4.0-6 Unit 2 - Amendment No. 196. I

4.0 2

Design Features Table 4.3.I<l*(pa{!e I of Ii For Fuel NOI Operated In Unils I and 2 Cycles I *4

. Coefficients to Calculale Ihe Minimum Required Fuel Assembly Burnup (Bu) as a Function of Decay Time and Enrichment (En)

FUEL DECAY TIME COEFFIC1 ENTS CATEGORY 2

o

-0.669 33.507 o

• 0.120

• H1.765 3

*. AJ*En2 + A~*En;- A.

Initial enrichment (En) is the nominal V-235 enrichmen\\. Any enrichment between 1.7 and 5.0 weight percent U*235 may be used. If the computed Su ~alue is negative. zcro shall be used.

3.

Decay Time is in years. 1\\0 assembly wiJh a cooling time greater than 20 years must use 20 years, No extrapolation is permil1ed.

4, If Decay Time "alue fall~ between increments of the table,the lower Decay Time value shall be used or a linear interpot;!tj~n ma)' be performed as foHows: Compute the Bu value using the coefficients associated with the Decay Time values thaI bracket the acltlaJ Decay Time. hllerpolale between Bu vulul."s based On the increment of Decay Time between the actual Decay Time value and the computed Bu results.

5.

Thi~ lable applies 10 fuel assemblies Ihat were nO! opernled in Ihe Unil I or Unit 2 core during operating Cycles 1 througl1 4.

Prairie Island Unit] - Amendment No. 209 UnitS 1 and 2 4.0-7 Unit 2 - Amendment No. 196

4.0 Design Features Any fresh fuel, irradiated fuel. or non* fuel material shall meeltl1e following restrictions prior to placement in the Spent Fuel Pool

,torage racks when any fuel IS in the spenl fuel pool:

A.

Any lIlTay of storage cells containing fucl shull comply with the storage paltems in Figure 4.3.1-' and [he requirements of Tables 4.3.1-L 4.3.1-2, and 4.3.1-3 as applicable. The category number o(fucl assemblies selected fora 2)(2 or 3x3 array (caregory dC-lcrmlned u,jng Table 4.3.1-2 or 43.1-3) shall be equal to ('r grealer (han the category number shown in the respeclive figure.

B:. Any siorage alTay location designated for a fuel assembly may be replaced with a failed fuel baskel (fuel rod siorage canister or failed fuel pin baskel), lncore detectors, or other non* fissile hardware.

C.

Fuel assembly insert8 designed for use in the fellC10T core may be inserled in a stored assembly (in the Spent Fuel POOl) wi1hOlit affecting the fuel catcg()ry.

Figure 4.3.1*1 (pa~e I ofJ)

Spent Fuel Pool Loading Resoiuions Prairie Island Unit 1-Amendment No. 209 Units 1 and 2 4.0-8 Unit 2 - Amendment No. 196

4.0 Design Fearures Array A Cacegory 6 assembly in eyery ctll Ctl<:ck~!rbc)ard pattern of diagonally-opposed Cucgory 1i!ssembJies with emplY cells.

x x

Array D Checkerboard panern of IINo face-adjacent Category.5 a5semb]je.~ wilh an emply cell and C~legory I aS1>embly, Allows for transition from Array C and other arrays, 5

Array E,

Chel:kcrooard pal1em uf Iwn diagonally.opposed Category 2 assemblies with an empcy cell and Calegory 4 assembly.

paHern of diagtmally-opposed Category 7 con~olidated rod slorage L'anisters and empty cells, which may be r.lled with assembly m.t;z1es. guide tubes. and grids.

x

,7 patlem of Category 5 assemblies wilh an RCCA Jo~ded in the cenler assembly, 5

5

.5

.5 5R 5

5 Figure 4,31*1 (page 2 of J)

Allowable SIOTa!!e Arrays prairie Island Unit 1 - Amendment No. 209' Units 1 and 2 4.0-9 Unit 2 - Amendment No:196 5

Design Features

4.0 NOles

I.. In all arrays, an assembly of higher Fuel Category l1umber can replace an assembly designated with a lower Fuel Categury number.

2.

Category I is fuel up 10 5.0 weight percent U-235 enrichment and does not credit bumup.

3.

Fuel Categories 2 through 6 Hre determined from Tables 4.3. J-2.or 4.3.1-3.

4.

An "R" designates a location.that requires inserTion of an RCCA in Ihe fuel assembly.

5.

An "X" designates a location that requires an empty cell, except that the empty cells in Array F may store assembly structural materials including nozzles, guide tubes, and grids.

6.

An*cmpty (water*filled) cell may lie substituted for any fuel-contail1ing cell in all stordge arrays.

7.

Array F shall only interface with Array A, and no other.

8.

Except for the center rudded a~sembly cif the 3x3 Array G and the special interface de filled between Array A and Array F, each assembly location is pari of up to f(lur 2x2 arrays (assembly in the lower right, lower left, upper right, upper lefl) and each Hssembly must simultaneously mectthe requiremenls of all those a.rrays of which it is a pan*.

Figure 4.3.1*1 (page 3 of 3)

Allowable Storage Arrays Prairie Island Unit 1 - Amendment No. 209

. Units 1 and 2 4.0-10 Unit 2 - Amendmem No. 196 I

OFFICIAL USE ONLY PROPRIETARY INFORMATION UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON. D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 209 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-42

.. AND AMENDMENT NO. 196 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-60 NORTHERN STATES POWER COMPANy - MINNESOTA PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 AND 2 DOCKET NOS. 50-282 AND 50*306

1.0 INTRODUCTION

By application dated A~gust 19. 2011 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML112360231), supplemented by tetters dated May 16. 2012.

September 4,2012, February 8,2013, and July 17, 2013 (ADAMS Accession Nos.

ML12139A198. ML12249A069, ML13039A306. and ML13199A416, respectively), Northern States Power Company, a Minnesota Corporation (the licensee), doing business as Xcel Energy, requested changes to the Technical Specifications (TSs) for Prairie Island Nuclear Generating Plant (P1NGP). Units 1 and 2.

The proposed changes correct non-cor:lservatisms in the PINGP spent fuel pool (SFP) criticality analysis-of-record, which the licensee has addressed in the PINGP Corrective Action Program.

The licensee has maintained operability of the SFP through interim administrative controls on SFP loading patterns that are more restrictive than the current TS requirements.

Specifically. the proposed changes would revise TS 3.7.17. "Spent Fuel Pool Storage: and TS 4.3.1, "Fuel Storage Criticality* to'provide new SFP loading restrictions that meet subcriticality for all postulated conditions. The amendments would also revise the evaluation methodology used for the SFP criticality analysis.

The licensee proposed seven different storage configurations and seven different fuel categories. Each fuel category, except for the category for the fresh 5 weight percent 235U fuel has a burnup versus initial enrichment requirement that must be met for safe storage. The storage configurations must also comply with acceptable interface requirements. In its application, the licensee submitted Westinghouse Report, WCAP-17400-P Revision 0, documenting PINGP's SFP criticality analysis. The proposed changes to TS 3.7.17, "Spent Fuel Pool Storage" and TS 4.3.1, "Fuel Storage Criticality" impose the storage requirements reflecting the new.SFP criticality analysis.

OFFICIAll:JSE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARVINFORMATION

- 2

.. The supplemental letters dated May 16, 2012, September 4,2012. February 8,2013, and July 17, 2013, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the Nuclear Regulatory Commission (NRC) staffs original proposed no significant hazards consideration determination as published in the Federal Register on February 14, 2012 (77 FR 8291).

Portions of the licensee's August 19, 20H, May 16, 2012, and September 4, 2012, submittals

. contain proprietary information and are therefore, withheld from public disclosure.

2.0 REGULATORY EVALUATION

The Commission's regulatory requirements related to the content of the TSs are contained in 10 CFR 50.36. The TS requirements in 10 CFR 50.36 include the following categories: (1) safety limits, limiting safety system settings, and limiting control settings, (2) limiting conditions for operation, (3) surveillance requirements, (4) design features, and (5) administrative controls.

The requirements for system operability during movement of irradiated fuel are included in the TSs in accordance with 10 CFR 50.36(c)(2), Limiting Conditions for Operation. As required by

• 10 CFR 50.36(c)(4), design features to be included are those features of the facility such as materials of construction and geometric arrangements, which, if altered or modified, would have a significant effect on safety and are not covered in categories described in paragraphs (c)(1),

(2), and (3) of 10 CFR 50.36. This amendment request concerns 10 CFR 50.36(c)(2).

10 CFR 50.36(c)(3), and 10 CFR 50.36(c)(4).

Paragraph 50.68(b)(1) of 10 CFR requires, "Plant procedures shall prohibit the handling and

. storage at anyone time of more fuel assemblies than have been determined to be safely subcritical under the most adverse moderation conditions feasible by unborated water."

Paragraph 50.68(b)(2) of 10 CFR requires, "The estimated ratio of neutron production to neutron

. absorption and leakage (k-effective) of the fresh fuel in the fresh fuel storage racks shall be calculated assuming the racks are loaded with fuel of the maximum fuel assembly reactivity and flooded with unborated water and must not exceed 0.95, at a 95 percent probability, 95 percent confidence level. This evaluation need not be performed if administrative controls and/or design features prevent such flooding or if fresh fuel storage racks are not used."

Paragraph 50.68(b)(3) of 10 CFR requires, "If optimum moderation of fresh fuel in the fresh fuel storage racks occurs when the racks are assumed to be loaded with fuel of the maximum fuel assembly reactivity and filled with low-density hydrogenous flUid, the k-effective corresponding to this optimum moderation must not exceed 0.98. at a 95 percent probability, 95 percent confidence level. This eyaluation need not be performed if administrative controls and/or design features prevent such moderation or if fresh fuel storage racks are not used.':

Paragraph 50.68(b)(4) of 10 CFR requires, "If no credit for soluble boron is taken, the k-effective of the spent fuel storage racks loaded with fuel of the maximum fuel assembly reactivity must not exceed 0.95. at a 95 percent probability. 95 percent confidence level, if flooded with unborated

• water. If credit is taken for soluble boron, the k-effective of the spent fuel storage racks loaded with fuel of the maximum fuel assembly reactivity must not exceed 0.95, at a 95 percent probability. 95 percent confidence level. if flooded with borated water, and the k-effective must remain below 1.0 (subcritical), at a 95 percent probability. 95 percent confidence level, if flooded with unborated water:

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- 3 Paragraph 10 CFR 50.S8{c) stipulates that, "While a spent fuel transportation package approved under Part 71 of this chapter or spent fuel storage cask approved under Part 72 of this chapter is in the spent fuel pool: (1) The requirements in § 50.SS{b) do not apply to the fuel located within that package or cask; and (2) The requirements in Part 71 or 72 of this chapter, as applicable, and the requirements of the Certificate of Compliance for that package or cask, apply to the fuel within that package or cask:

)

The U.S. Atomic Energy Commission (AEC>"issued its Safety Evaluation (SE) for PINGP before the revised General Design Criteria (GDCs) were published in 1971. A PINGP GDC requires,

that, "Criticality in new and spent fuel storage shall be prevented by physical systems or processes. Such means as geometrically safe configurations shall be emphasized over procedural controls."

On September 29. 2011, the NRC staff issued the Interim Staff Guidance (lSG)

DSS-ISG-2010-01 (ADAMS Accession No. ML110620086). The purpose of the ISG is to provide updated review guidance to the NRC staff to address the increased complexity of recent SFP nuclear criticality analyses and operations. The NRC staff used ISG OSS-ISG-2010-01 for the review of the current application.

As guidance for reviewing criticality analyses of fuel storage at light-water reactor power plants.

the NRC staff issued an internal memorandum on August 19, 1998 (ADAMS Accession No. ML00372B001).. This memorandum is known as the "Kopp Letter,P after the author, Laurence Kopp. The Kopp Letter provides guidance on s.alient aspects of a criticality analysis. The guidance is germane to boiling-water reactors and pressurized-water reactors, and'to borated and unborated conditions.

Additional guidance is available in NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light~Water Reactor] Edition,"

particularly Section 9.1.1, "Criticality Safety of. Fresh and Spent Fuel Storage and Handling,"

Revision 3, issued March 2007. Section 9,1.1 provides the existing recommendatiqnsfor performing the review of the nuclear criticality safety analySiS of SFPs.

The NRC staff used the following. regulatory guidance in reviewing the human performance aspects of the license amendment request:

NUREG-1764, "GlJidance for the Review of Changes to Human Actjons,~ provides guidance to determine the appropriate level of human factors engineering (HFE) reView of human actions credited for safety, based on their risk importance.

NUREG-0711, KHuman Factors Engineering Program Review Model," Revision 2, was used by the NRC. staff to review licensees' HFE programs to verify that these programs incorporate HFE practices and guidance accepted by the staff.

NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor] Edition" Chapter 18.0, "Human Factors Engineering,*

Revision 1, provides guidance to the NR.G staff in the review of the HFE aspects of modifications affecting risk-important human actions.

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• 4

3.0 TECHNICAL EVALUATION

3.1 Proposed Changes The licensee proposed seven different storage configurations and seven different fuel categories. Each fuel category, except for the category for the fresh 5 weight percent 23SU fuel, has a burnup versus initial enrichment requirement that must be met for safe storage. The storage configurations must also comply with the interface requirements. Th~ proposed changes to TS 3.7.17, "'Spent Fuel Pool Storage" and TS 4.3.1, "Fuel Storage Criticality" impose the storage requirements reflecting the new SFP criticality analysis. Specific changes have also been proposed to im,prove the TS structure and expand the scope to include the placement of fuel inserts and other hardware that may affect criticality.

3.1.1 Changes to TS 3.7.17, "Spent Fuel Pool Storage" The proposed change will remove TS Figure 3.7.17 -1, which has been used to discriminate between restricted and non-restricted locations in the SFP. The licensee proposed to refer personnel to TS 4.3.1.1 for all SFP loading restrictions, thereby eliminating the concept of "restricted" and "unrestricted" loading in the SFP (Le., all loading will be "restricted"). In addition, the proposed TS 3.7.17 will expand the LCO and Surveillance Requirement (SR) 3.7.17.1 to address the storage of fuel inserts and hardware.

3,1.2 Changes to TS 4.3.1, "Fuel Storage Criticality" Regarding the proposed change to TS 4.3.1.1.c, the value-of SFP b9ron concentration that ensures that conditions in the SFP maintain a neutron multiplication factor (keff) less than or equal to 0.95 will be revised. Additionally, the proposed change will modify sub-item (e) to apply.

the new loading requirements of TS Figure 4.3.1-1 and expand the scope to include fuel inserts and hardware. The proposed change will replace current loading restrictions in TS Figures 4.3.1-1 through 4.3.1-4 with new loading restrictions embodied by TS Tables 4.3.1-1 through 4.3.1-3 and T5 Figure 4.3.1-1.

The proposed change will also delete TS 4.3.1.3 in its entirety with no replacement. Currently, this TS has imposed a 10 CFR Part 72 spent fuel cask loading restriction that is more appropriately addressed in PINGP Independent Spent Fuel Storage Installation TSs (Special Nuclear Material license SNM-2506). The proposed Figure 4.3.1-1 makes provisions for spent fuel pool contents that were not previously included in the T5 These contents include conSOlidated rod storage canisters, failed fuel baskets~ and fuel assembly inserts.

3.2 Reactor Systems 3.2.1 Method of Review With its application, the licensee submitted Westinghouse Report. WCAP-17400-P Revi~jon 0, documenting PINGP's criticality analysis. The NRC staffs review was performed consistent with Section 9.1.1 of NUREG-0800.

On September 29, 2011, the NRC staff issued the Interim Staff Guidance (ISG)

DSS-ISG-2010-01 (ADAMS Accession No. ML110620086). The purpose of the ISG is to provide updated review guidance to the NRC staff to address the increased complexity of recent OFFICIAL USE ONLY PROPRIETARY INFORMA.TION

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- 5 SFP nuclear criticality analyses and operations. The NRC staff used ISG DSS-ISG-201 0-1 for the review of the current application.

On August 19, 19~8, the NRC staff issued an internal memorandum containing guidance for reviewing criticality analyses of fuel storage at light-water reactor power plants. This memorandum is known colloquially as the "Kopp Letter" (ADAMS Accession No. ML003728001),

after the author, Laurence Kopp. While the Kopp Letter does not specify a methodology, it does provide some guidance on the more salient aspects of a nuclear criticality safety (NCS) analysis,

. including computer code validation. The guidance is germane to boiling-water reactors and pressurized-water reactors (PWRs), and to borated and unborated conditions. The NRC staff used the Kopp Letter for the review of the current application.

3.2.2 SFP NCS Analysis Review 3.2.2.1 Computational Methods For the criticality calculation, the licensee used SCALE Version 5.1, with the 44 group Evaluated Nuclear Data File, Version 5 (ENDF/B-V) neutron cross section library. SCALE is a comprehensive modeling and simulation suite for nuclear safety analysis* and design. developed and maintained by Oak Ridge National Laboratory under contract with NRC and the U.S.

Department of Energy (DOE) to perform reactor physics, criticality safety. radiation shielding, and spent fuel characterization for nuclear facilities and transportation/storage package designs.

For the depletion calculation to determine the spent fuel isotopics, the licensee used the two-dimensional PARAGON code with an Evaluated Nuclear Data File, Version 6 (ENDF/B-VI) neutron cross section library. PARAGON has been approved by the NRC for depletion analysis (ADAMS Accession No. ML042250311). These computer codes and the nuclear data sets with them have been used in many NCS analyses, and are industry standards. Therefore, the NRC staff considers their use in the current application to be acceptable.

3.2.2.2 Criticality Code Validation The purpose of the criticality code validation is to ensure that appropriate* code bias and bias uncertainty are determined for use in the criticality calculation. The ISG DSS-ISG-2010-01 references NUREG/CR-6698, "Guide for Validation of Nuclear Criticality Safety Calculational Methodology."

I NUREG/CR-6698 states. in part, that:

In general, the critical experiments selected for inclusion in the validation must be representative of the types of materials, conditions, and *operating parameters found. in the actual operations to be modeled using the calculational method. A sufficient number of experiments with varying experimental parameters should be selected for inclusion in the validation to ensure as wide an area of applicability as feasible and statistically significant r.esults.

The NRC staff used NUREG/CR-6698 as guidance for review of the code validation methodology presented in the application. T~e basic elements of validation are outlined in NUREG/CR-6698, including identification of operating conditions and parameter ranges to be validated, selection of critical benchmarks, modeling of benchmarks, statistical analysis of results,.and determination of the area of applicability.

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- 6 SCALE used in both the code benchmark analysis and the fuel storage analysis, includes the control module CSAS25 and the following functional modules: BONAMI, NITAWL-III, and KENO V.a (KENO). The licensee performed the validation of the CSAS25 sequence by comparing KENO calculated k-effective values with several different sets of critical configurations. A total of

((

))1 critical configurations were included; The licensee determined and applied separate sets of bias and bias uncertainty based on the specific stqrage conditions. The sources of critical configurations ((

))

((

]J Therefore, the NRC staff considers ([

J] to be an appropriate source of information for the critical..

experiment models. Critical experiments ((

)) contain important features such as soluble boron and poisoned fuel rods. ((

)) The NRC staff has reviewed the experiments used in the validation of the critiCality code for the PINGP SFP and considers them appropriate for that use.

Fission product k-effective validation was identified by the applicant as a validation gap. The analysis uses ((

lJ of the fission product worth as an uncertainty to cover the fission product validation gap.. The value is based on the licensee's engineering judgment of fission product nUcl.ear data. The NRC staff concludes that the approach used to determine the uncertainty and the derived values is appropriate ((

lJ*

The licensee identified the applicable operating conditions for the validation (e.g., fissile isotope, enrichment of fissile isotope, fuel chemical form, types of neutron absorbers, moderators and reflectors, range of moderator to fissile isotope, and physical configurations). ((

)) The licensee performed a trend analysis and identified an additional bias for those parameters with a statistically significant trend..

Based on the above discussion, the NRC staff concludes that the information supporting the code validation is acceptable.

I The use of double box parentheses identifies the enclosed information as proprietary.

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3.2.2.3 Fuel-Assembly Selection Section 3.1 of WCAP-17400-P provides information on fuel assembly selection. The licensee analyzed current, future, and all legacy fuel assembly designs used or expected to be used at PINGP to establish the limiting fuel assembly design. Westinghouse supplied the STANDARD (STD) fuel assembly for Cycles 1 through 4 for both units. Exxon Nuclear Fuel (ENC) supplied three different fuel assembly designs, ENC STANDARD, ENC HIGH BURNUP, and ENC TOPROD for cycles 5 through 10. Beginning with Cycle 11, Westinghouse has supplied the Optimized Fuel Assembly (0FA), HIGH BURNUPOFA, 400 VANTAGE+ (400V+), and 422 VANTAGE+ (422V+) fuel assembly designs.

The licensee performed a fuel assembly reactivity comparison and concluded that the 422V+

fuel assembly design, assuming a bounding theoretical density (TD), is an acceptable reference fuel assembly design for subsequent criticality analysis. The licensee depleted the OFA. STD.

and 422V+ fuel assembly designs covering the applicable burnup range. The licensee used the resulting isotopic information in KENO to determine the reactivity of each fuel assembly design.

A comparison showed that the reactivity of STD and 422V+ was comparable throughout the burnup range. The NRC staff noted that the comparison was performed using only a single array configuration, ((

)).

The active fuel of the STD fuel assembly is 0.75 inches longer and uses Zircaioy-4 as the cladding material. The.422V+ fuel assembly uses the ZIRLO 'material. The licensee conservatively modeled the less neutron-absorbing cladding material to maximize reactivity.

The reactivity of OF A was ((

)).

The ENe fuel is Significantly less reactive compared to the 422V+ fuel. The licensee determined that the ENC fuel assembly designs are about BOO to ~OOO pcm less reactive compared to the 422V+ depending on the burnup. The three ENC fuel types have a stack density of ([

J] TD as compared to the II

)) TD assumed for 422V+.

Based on the noted design considerations, the NRC staff considers acceptable the selection of 422V+ as the reference assembly design.

3.2.2.4 Depletion Analysis

• Section 3,3 of WCAP-17400-P provides information on the depletion analysis, To take credit for the reduction in reactiyitydue to fuel burnup, the spent fuel composition should be based on an OFFICIAL USE ONLY PROPRIETARY INFORMATION

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-8 appropriate depletion analysis with proper assumptions regarding depletion uncertainty.

depletion parameters, and axial burnup and temperature profiles.

3.2.2.4.1 Depletion Uncertainty The licensee used the two-dimensional PARAGON code to calculate the isotopic composition of the spent fuel as a function of fuel burnup. initial feed enrichment, and decay time. The NRC staff has approved PARAGON for PWR depletion calculations as a part of its approval of "Qualification of the Two-Dimensional Transport Code PARAGON: WCAP*16045*P-A (ADAMS Accession No. ML04225031.1). The uncertainty in the kef! introduced by the dep!etion isotopiC uncertainty was addressed by applying ((

)) as an uncertainty component in the determination of the maximum kef(. The NRC staff concludes that this uncertainty treatment is a'cceptable because it is consistent with ISG DSS-ISG-2010-01.

3.2.2.4.2 Depletion Parameters The ISGDSS-ISG-2010-01 provides guidance that depletion simulations should be performed with parameters that maximize the reactivity of the depleted fuel assembly. The ISG DSS-ISG 2010-01 references NUREG/CR-6665, "Review and Prioritization of Technical Issues Related to Burnup Credit for LWR Fuel'," (Reference 5) which discusses ~he treatment of depletion parameters. For fuel and moderator temperatures, NUREG/CR-6665 recommends using the maximum operating temperatures to maximize plutonium production.

The NUREG/CR-6665 also recommends using a conselVative cycle average boron concentration for the depletion analysis. The licensee used a'boron concentration that bounds past and anticipated future cycle average boron concentrations for bot,h units, The NUREG/CR*6665 does not have a specific recommendation for specific power and operating history. The NUREG/CR-6665 estimated this effect to be about 0.002llkeff using operating histories it considered. Based on the difficulty of reproducing a bounding or even a representative power operating history, NUREG/CR-6665 merely recommends using a constant power level and retaining sufficient margin to cover the potential effect of a more limiting power history. The licensee used a [I J1 for the depletion calculations and applied a ((

nuncertainty on the operating history.

To date. PINGP has used two different types of burnable poisons: BUrnable Poison Rod Assemblies (SPRAs). and gadolinia integrated fuel rods. The licensee confirmed that no other reactivity control device has been used at PfNGP. The licensee stated that the BPRA bearing assemblies were only used in the first four cycles of each unit and these assemblies were enriched to a maximum of 3.4 weight percent 235U. For these assemblies, the licensee,

determined the isotopic number densities assuming conselVative conditions and a bounding

, BPRAJoading to maximize the reactivity of the discharged fuel, The licensee calculated a '

separate set of burnup versus enrichment CUlVes for these assemblies used in the first four cycles.

The licensee stated that the current and future approach for power distribution control at PINGP is to use gadolinia which is included as an integral part of the fuel matrix. The NUREG/CR, 6760, "Study of the Effect of Integral Burnable Absorbers for PWR Bu(nup Credit," showed that throughout burnup, the reactivity for fuel assemblies containing gadolinia remain lower than the reactivity for fuel assemblies without gadolinia due to the residual poison that will not completely OFFICIAL blSE ONlY PROPRIETARY INFORMATION

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- 9 burn out of an assembly. Based on this result. the licensee concluded that not crediting gadolinia is conservative and modeled the fuel accordingly. The NRC staff noted that the referenced NUREG-6760 analysis was based on a cask analysis with poison plates and the corresponding conclusions mayor may not be applicable to the PINGP spent fuel pool analysis which assumed un-poisoned racks. The licensee provided site-specific analysis that, given the assumptions made in that analysis; demonstrates that the PINGP method of addressing the effects of gadolinia in its fuel is acceptably conservative.

The licensee considered the impact of rodded operation as it affects the reactivity of the discharged assembly. The licensee elected to treat the assemblies that experienced rodded operation in two parts. For fuel assemblies from Cycles 1 through 4, the licensee concluded that depletion with burnable poison fully inserted in the BPRAs would bound depletion with partially' inserted RCCAs. The NRC staff finds this reasonable since duration of rodded operation would be bounded by the depletion with BPRA over the life of the fuel. For fuel assemblies from Cycle 5 to the present and into the future. the licensee claimed that up ~o 1 GWd/MTU of rodded operation would be bounded by the design basis analysis used to develop the burnup requirements. In response to NRC staff RAls,. the licensee provided quantitative information to.

support the proposed design basis loading curves.

Rodded Operation In its February 8.2013, 'Ietter, the licensee proposed the following commitment:

In conjunction with implementation of the proposed TS, procedures will be revised to require an assessment of a fuel assembly's exposure to rodded power operation in the core prior to moving that fuel assembly into the spent fuel pool (SFP) storage racks. If an assembly experiences more than 100 megawatt day per metric ton uranium (MWdIMTU) of core*

average full-power rodded operation exposure in the cycle immediately prior to discharge to the spent fuel pool, this exposure experienced while rodded will not be credited for determining the coefficients used to categorize fuel assemblies as described in WCAP 17400-P. In addition, if an assem~ly experiences more than 1 gigawatt day per metric ton uranium (GWd/MTU) of core averagerodded operation lifetime exposure. the assembly shall be either treated as Fuel Category 1 or evaluated to determine which Fuel Category is appropriate for safe storage of the assembly.

In its July 17, 2013. letter. the licensee proposed the following reVisions to the commitment:

Striking the phrase "in the cycle immediately prior to discharge to the spent fuel poo!."

Striking the last sentence of the commitment With these changes the commitment now reads as follows:

In conjunction with implementation of the proposed TS. procedures will be revised to require an assessment of a fuel assembly's exposure to rodded power operation in the core prior to moving that fuel assembly into the spent fuel pool (SFP) storage racks. If an assembly experiences more than 100 megawatt days per metric ton uranium (MWd/MTU) of core average full-power rodded operation exposure, this exposure experienced while rodded will not be credited for determining the coefficients used to categorize fuel asserl)blies as described in WCAP*17400-P.

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- 10 Rodded operation will affect the reactivity of discharged spent nuclear fuel (SNF) in several ways..The thermal neutron absorption in the control rods will harden the neutron spectrum resulting in increased production of fissile plutonium isotopes. It will also skew the axial burnup profile. These effects would raise the SNF's net reactivity. However, the fuel assembly would also likely see reduced moderator and fuel temperatures, effects that would reduce the SNF's net reactivity. The licensee has rod insertion limits that allow them several options to operate at power with rods inserted. The NRC staff did not believe the licensee's analysis fully addressed "

the variability that would be required to estimate the reactivity of SNF that had experienced significant amounts of rodded operation. The current version of the commitment does not allow for significant amounts of rodded operation to be considered when determining whether a fuel assembly meets the storage requirements. Therefore, the NRC concludes that this commitment acceptably accounts for the variability.of rodded operation.

A requirement is included in Section 3, "Implementation Requirements," of the NRC license amendments associated with this s"afety evaluation, in order to ensure that the proposed procedure changes in the above commitment are incorporated coincident with the licensee's implementation of the amendments.

3.2.2.4.3 Axial 8urnup and Temperature Profiles At the beginning of life. a PWR fuel assembly will be exposed to a near-cosine axial-shaped flux, which will deplete fuel near the axial center at a greater rate tAan at the ends. As the' reactor continues to operate. ttie cosine flux shape wlll flatten because of the fuel depletion and fission-product buildup that occurs near the center. Near the fuel assembly ends, burn up is.

suppressed due to leakage. If a uniform axial burn up profile is* assumed. then the burnup at the ends is over predicted. Analysis discussed in NUREG/CR-6801, "Recommendations for Addressing Axial Burnup in PWR 8urnup Credit Analyses," has shown that, at assembly burnups above about 10 to 20 GWd/MTU, this results inan underprediction of k-eflective.

Generally the underprediction becomes larger as burnup increases. This is what is known as the "end effece Proper selection of the axial burn up profile is necessary to ensure k-effective is not underpredicted due to the end effect.

The NUREG/CR-6801 provides recommendations for selecting an appropriate axial bumup profile. The NCS analysis documented in WCAP-174oo-p did not use the axial burnup profiles from NUREG/CR*6801. A description of how the axial burnup profiles were derived is provided in Section 3.3.3 of WCAP-17400-P. The licensee used ((

)) axial burnup profiles from past and current core designs to derive five bumup profiles corresponding to ((

)].

When the analysis was initially performed, it included the anticipation of a future extended power uprate (EPU) license amendment request (LAR) and several aspects of the analysis were intended to bound that EPU. However, in its February 8, 2013, letter (ADAMS Accession No. ML13039A306). the licensee indicated that it no longer intended to submit an EPU LAR and the issue regarding the'EPU's effect on post irradiated fuel has not been resolved. Additionally. the licensee may at some tirne in the future reconsider its decision and submit a power uprate LAR.

Therefore, any effects of a future power increase are not fully covered by this license amendment and would need to be addressed at that time. From the database of profiles*.the method identifies ([

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- 11 U. The NRC staff requested additional information regarding the derivation of the axial burnup profiles. Based on the licensee's response, the NRC staff determined that the method selects the axial burnup profile corresponding to a spent fuel assembly with the most underburned top region ofthe fuel.

Therefore, the method is accept~bly conservative.

In its response to an NRC staff RAI. the licensee also showed that the use of annular pellets in the axial blankets,((

1],

Based 'on the above discussion, the NRC staff concludes that the licensee selected an appropriate bumup profile.

3.2.3 Criticality Analysis 3.2.3.1 Normal Conditions The PINGP Units 1 and 2 share a common spent fuel storage pool that employs one modular storage rack design throughout t~e pool. The PINGP utilizes the high density racks that were designed with Boraflex neutron absorber inserts. The criticality analysis does not credit any remaining Boraflex neutron absorber material. The model assumes that the remaining Boraflex material ((

)). The criticality analysis is based on ((

)). During normal operation, at least 1800 ppm of soluble boron is present in the SFP and the moderator temperature is less than or equal to 150 of. The licensee proposes seven normal storage configurations for use throughout the PINGP spent fuel pool.

The manufacturing tolerances of the storage racks and fuel assemblies contribute to the reactivity. Consistent with the Kopp Letter, the determination of the maximum k-effective should consider either: (1) a worst-case combination with mechanical and material conditions set to maximize k-effective, or (2) a sensitivity study of the reactivity effects of tolerance. variations. if used, a sensitivity study should include all possible significant tolerance variations in the material and mechanical specifications,of the racKs.

The licensee's analysis evaluated the following uncertainty components: ((

)). The NRC staff found that this assumption was acceptable in section 3.2.2.3 above, To determine the reactivity uncertainty associated with a specific manufacturing tolerance, the licensee used KENO to calculate the k-effective for the reference condition and th,e k-effective for the perturbed case. The reference condition is the condition with nominal dimensions and p*roperties. All tolerance perturbations were applied in the qirection that increases reactivity relative to the nominal condition. ((

)) Based on the considerations discussed above, the NRC staff finds the licensee's treatment of manufacturing tolerances acceptable be.cause it is consistent with the guidance in the Kopp Letter.

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- 12 According to the Kopp Letter, the criticality analysis should account for the temperature corresponding to the highest reactivity. The licensee performed the base criticality analysis at

((

Jl To address the full range of allowable operating temperatures, the licensee performed additional KENO calculations ((

J]. The maximum positive difference was applied as a bias to the maximum k*

effective.

The NRC staff concludes that this is consistent With the Kopp Letter, and, therefore. acceptable.

3.2.3.2 Abnormal Conditions Section 4.5 of WCAP-17400-P presents the abnormal conditions considered in the analysis.

The licensee considered the following abnormal conditions:

misloaded assembly

.* inadvertent removal of an RCCA spent fuel temperature outside of operating range, and dropped and misplaced fuel assembly The licensee determined that the limiting abnormal condition was the misloaded assembly. The licensee created models for.the misloaded assembly by placing a fresh 5.0 weight percent fuel assembly into a water-filled cell for the storage configurations requiring water filled cells (Arrays B. C, D. and E). The licensee's analysis determined that the limiting Array ((

)) would require 910 parts per million (ppm) of soluble boron.to comply with the regulatory k~effective limit of 0.95.

The licensee did not explicitly calculate the k-effective for the inadVertent removal of an RCCA from Array G claiming that the misload event is bounding. Removal of an RCCA from Array G results in an array uniformly loaded with Category 5 fuel assembly. The misload event results in an array with two fresh assemblies (Category 1) face adjacent and two Category 5 fuel assemblies face adjacent in a 4x4 array. Therefore. the NRC staff agrees with the licensee that the limiting misload event would bound the condition following an inadvertent RCCA withdrawal.

To cover the SFP heat up conditions, the licensee analyzed the pool with a moderator density of 0.96 grams per cubic centimeter, which corresponds to the lowest water density at boiling point*

at atmospheric conditions. In addition. the lice.nsee analyzed the pool with a moderator density of 0.75 grams per cubic centimeter. The k-effective results showed that both cases were bounded by the misload event.

The licensee stated that the dropped fuel assembly is non-limiting because of the separation between a dropped assembly lying across the top of the fuel assembly top nozzle and the active fuel region, would be less reactive than a misloaded fuel assembly.

The licensee stated that the misplacement of a fuel assembly alongside the racks is less reactive than the misloaded fuel assembly because the misplaced fuel assembly will be

. bordered on two sides by water, and the leakage present in that scenario would make for a less reactive arrangement than a misloaded' fuel assembly, which would have fuel on all sides.

The accidents considered are reasonable considering the guidance in the Kopp Letter and the margin available between the amounf of soluble'boron required to maintain k-effective less than or equal to 0.95 and the soluble boron concentration required by the technical specifications.

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-13 Based on the above discussion, the NRC staff finds the licensee's evaluation of the accident conditions acceptable.

3.2.4. Staffs Evaluation of Elimination of TS 4.3.1.3 from TSs TS 4.3.1.3 was added to the PINGP TSs by Amendment Nos. 99 and 92 for Unit 1 and Unit 2,

. respectively, on July 9, 1992 (ADAMS Accession No. ML022210492). This pre-dates the promulgation of 10 CFR 50.68, which was originally issued in November 1998. The NRC revised 10 CFR 50.68 in November 2006 to add paragraph 10 CFR 50.68{c). Paragraph 10 CFR 50.68{c) was added to address the' issue identified in NRC Regulatory Issue Summary

. 2005-05, "Regulatory Issues Regardil')g Criticality Analyses for Spent Fuel Pools and Independent Spent Fuel Storage Installations."

With the issuance of the November 2006 revision to 10 CFR 50.68, there was no longer a need for licensees to perform separate nuclear criticality safety analyses for a cask when it was in the spent fuel pool. Since the November 2006 revision to 10 CFR 50.68, the nuclear criticality safety requirements for a cask are governed by the applicable Part 71 and/or Part 72 license.

Therefore, the inclusion of cask loading nuclear criticality safety requirements in the Part 50.

license is duplication and creates the possibility of an inadvertent violation if the Part 50 license is not revised when the Part 71 and/or Part 72 license is revised. Therefore, the NRC staff finds it acceptable to delete TS 4.3.1.3 in its entirety.

3.2.5 Summary The licensee submitted an LARto address the non-conselVatisms in the spent fuel pool nuclear criticality safety analysis of record and the associated TSs. The LAR is supported by Westinghouse Report, WCAP-17400-P, Revision 0, which documents the criticality analysis for PtNGP spent fuel storage. The proposed changes to TS 3.7.17, "Spent Fuel Pool Storage" and TS 4.3.1, "Fuel Storage Criticality" impose the storage requirements reflecting the new SFP criticality analysis..

The NRC staff reviewed the analysis to ensure that the assumptions and analytical technique used are adequately substantiated to conclude at a 95 percent probability, 95 percent confidence level, that the regulatory requirements will be met.

Based on the discussions in Section 3.2 of this safety evaluation, thEl NRC staff finds reasonable assurance that PINGP will comply with the applicable regulatory requirements. Therefore, the NRC staff concludes that the proposed TS changes are acceptable:

3.3 Human Performance and Health Physics 3.3.1 Potential Effects on Human Performance In its application, the licensee stated that.... the proposed amendments involve no phYSical modifications to the SFP storage racks orto any other system, structure, or component. No change to the minimum SFP boron concentration limit is required. The only physical effect associate~ with this proposed amendment will be the reconfiguration of fuel in the SFP storage racks.~ With respect to the possibility of increased risk due to human errors. the NRC staff '

reviewed the potential impact of the proposed changes to TS loading restrictions on OFFICIAl USE ONlY PROPRIETARY INFORMATJON

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- 14.

opportunities for error,and the'current of proposed barriers in place to prevent or mitigate human errors,

, 3.3.1.1 Changes to Design Basis Operator Actions The NRC staff has determined that the proposed changes do not impact the operator actions or their timing as credited in the design basis accidents analyzed in Section 14.5.1 "Fuel Handling,"

of the PINGP Updated Safety Analysis Report (USAR) and that it is appropriate to continue to credit these actions.

3.3.1.2 Changes to Fuel Characterization Regarding the actions involved in characterizing fuel, the licensee stated that the only change is the use of Core Operating Cycle instead of gadolinium content to distinguish which TS table applies to a specific irradiated fuel.assembly, Because both of these parameters are obtained from procurement or operating records, the actions associated with these two parameters (Core Operating Cycle and gadolinium pontent) are essentially the same. From the human factors standpoint, the risk.associated with identifying the wrong Core Operating Cycle is equivalent to the current risk of mistakenly. identifying a non-gadolinium assembly as a gadolinium assembly.

Therefore, the NRC staff concludesthatthere is no appreciable increase in the probability of human error due to this proposed change. The following barriers to prevent characterization errors are currently used and will continue to be used after the proposed changes are implemented:

• Training of nuclear engineers and operators Qualification of nuclear eng'ineers and operators Documentation of fuel assembly characteristics (availability for use in the characterization process) 3.3.1.3 Changes to Fuel Categorization A fuel assembly's category determines which storage arrays are acceptable for which assemblies. The process for categorization is being changed to make it more consistent.

Previously, fuel assemblies were first screened as "restricted" or:non-restricted", Now, this pre-screening is not necessary - all assemblies are considered "restricted" and therefore, go through a thorough categorization process based on reactivity. New categories have been created to 'cover everything from the highest reactivity (Category 1) to lowest (Category 6) and an additional Category 7 for ConSOlidated Rod Storage Canisters. Once the "ShuffleWorks' software in use at PINGP has beeh updated with the definitions of the neiN categories, "ShuffleWorks" will categorize a fuel assembly based on its burn-up requirements. From this point, the process will remain the same as current practices. Therefore, the NRC staff concludes that there is no appreciable increase in the probability of human error due to this proposed change.

The following barriers to prevent categorization errors are currently used and will continue to be used after the proposed changes are implemented:

Experience of nuclear engineers and operators Qualification of nuclear engineers and operators

• Operating experience with "ShuffleWorks" software Independent verification of "ShuffieWorks" output (Fuel Transfer Logs (FTLs>>

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- 15 Independent categorization of affected assemblies and preparation Qf FTLs by qualified individuals 3.3.1.4 Configuration Control The licensee stated ttiatthe process of moving fuel is currently controlled by procedure and that the overall process will not change. That is, fuel mov'ers will continue the current practice of following FTLs.when moving and storing fuel assemblies. The nuclear engineers who develop the FTLs will be given one-time training in th~ proposed, new TS criticality requirements and storage confIguration requirements.

The only substantive change in confige.Jration control in the SFP is that cells that are identified as empty cells will be controlled like any other cell. Affected personne(will be notified that empty cells in the SFP must remain empty and a note will be included in the TS to prevent the previously acceptable practice of storing inconsequential, non-fuel hardware in empty cells.

There are multiple barriers in place to prevent mis-positioning of an assembly. None of these will change or be negatively impacted by the proposed change to the.TS. Therefore, the NRC staft,concludes that there i~ no appreciable increase in the probability of human error due to this proposEld change.

. Th.e following barriers to prevent fuel movement errors are currently used and wilt continue to be used after the proposed changes are implemented:

Approved procedures Independent verification of fuel categorizations Independent verification of FTLs Use of a Fuel Handling Supervisor or Fuel Accountability Engineer to verify "to" and "from" locations

• Movement by one step at a time

• Use of a step verifier

• Pre-job briefs

• Use of three-way communication

• Post-campaign inventory 3.3.2 Summary Based on the discussions in Section 3.3 of this safety evaluation. the NRC staff concludes that the proposed TS changes do not affect manual actions credited in PINGP USAR, Chapter 14, Section 14.5.1. "Fuel Handling." The staff further concludes that the proposed TS changes are acceptable from the human performance perspective based on the licensee's statements that appropriate training will be provided, procedures and other administrative controls will continue to be applied. qualified nuclear engineers will be used to categorize fuel assemblies and to identify proper storage locations, and fuel movement equipment. including controls, displays, anq alarms. will not be affected.

3.4 Licensina Basis Change for SFP Criticality In its August 19, 2011 \\ application, the licen,see requested NRC approval to change its regulatory basis for SFP criticality analysis from 10 CFR 70.24 to 10 CFR 50.q8(b), which would allow for elimination of criticality accident monitoring requirements, while maintaining the subcriticality criteria defined by 10 CFR, 50.68(b). The regulations at 10 CFR 50.68(a} state that OFFICIAL U8E ONLY PROPRIETARY INFORMATION

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- 16 the licensee shall comply with either 10 CFR 70.24 or 10 CFR 50.68(b). Therefore, the licensee may comply with 10 CFR 50.68(a) by complying with 10 CFR 50.66{b) in lieu of maintaining a monitoring system capable of detecting a criticality as described in 10 CFR 70.24. The licensee

,does not require prior NRC approval to comply with 10 CFR 50.68(a) in this manner.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Minnesota State official was notified of the proposed issuance of the amendments. The State official had' no comments.

5.0 ENVIRONMENTAL CONSIDERATION

'The 'amendments change the requirements with respect to the installation or use,of a facility component located within the restricted area as defined in 10 CFR Part 20 or change the surveillance requirements. The NRC staff has determined that the amendments involve no significant increase in the a'mounts, and no significant change in the types, of any effluents that

, may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration and there has been no public comment on such finding (77 FR 8291). Accordingly, the,amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9), Pursuant to, 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.

6.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above. that: (1) there is reasonable assurance that the health and safety of the public will not be endangereq by operation in the proposed manner, (2) there is reasonable assurance that such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributors: K. Wood T:Nakanishi G: Lapinsky Date: August 29, 2013 OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMA"RON J. lynch

- 2 The NRC has determined that the related Safety Evaluation (SE) contains proprietary information pursuant to Title 10 of the Code of Federal Regulations, Section 2.390, 'Public Inspections, Exemptions, Requests for Withholding." Proprietary information is in~icated by text enclosed within double brackets. Accordingly, the NRC staff has also prepared a redacted publicly available, non-proprietary ver~ion of the. SE. Copies of the proprietary and non proprietary versions of the SE are enclosed. The Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Sincerely,

,./RN '

Thomas J. Wengert, Senior Project Manager Plant licensing Branch 111-1.

Diyision of Operating Reactor licensing Office of Nuclear Reactor Regulation Docket Nos. 50-282 and 50-306

Enclosures:

1. Amendment No. 209 to DPR:-42

2. Amendment No. 196 to DPR:-60 3., Non-Proprietary Safety Evaluation

4. Proprietary Safety Evaluation cc w/encls 1, 2, and 3: Distribution via listServ DISTRIBUTION:

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